In a pressurized-water reactor the fuel elements are arranged in parallel with one another inside the core of the reactor, and for the control of the reactor control rods of material which absorbs the neutrons are moved between these fuel elements. The fuel elements are immersed in the primary fluid which is water under pressure, and which serves both as coolant fluid and moderator. The fuel elements delimit the coolant channels between them.
During operation of the reactor it is necessary to avoid damaging the fuel elements which can undergo an irreversible degradation under the effect of the heat which they release, if this amount of heat is too large to be absorbed and carried away by the coolant, or if the flow of the coolant is insufficient.
In the control of a pressurized-water nuclear reactor, it is necessary to provide against two different phenomena which are accompanied by a more or less complete destruction of the fuel. These two phenomena are, on the one hand, the crisis of boiling of the water under pressure in contact with the fuel elements, and on the other hand, the melting of the fuel.
In short, at the time of exchange of heat between the fuel and the water under pressure across the outer surface of the fuel rods the water, although maintained under high pressure, reaches under the effect of the high temperature of the fuel elements, a state close to boiling. A limited boiling phenomenon in the vicinity of the fuel elements may even favor the thermal exchanges between the fuel elements and the water under pressure. It is necessary, however, to prevent this boiling from leading to the formation of a film of steam along the fuel elements, because such a film of steam prevents large exchanges between the fuel element and the water under pressure, because of calefaction.
Hence it is extremely important to watch for the appearance of the starting of boiling with the production of steam in contact with the fuel elements, or the crisis of boiling, in order to avoid, during operation of the reactor, entry into a zone of operation where this crisis of boiling appears.
Up to the present, an attempt has always been made to operate under conditions providing a margin of safety between the conditions of operation of the reactor and the conditions in which the crisis of boiling would appear, in order to take into account the accuracy of the measurements and of the methods of evaluation of the conditions of appearance of the crisis of boiling.
On the other hand, a second type of phenomenon which can appear at the level of the fuel elements and lead to their destruction is the melting of one or more pellets of the fuel elements under the effect of too great a release of power which appears as a local rise in temperature to the melting point of the material forming the fuel pellets, this release of power being able to lead to melting when its value per unit length or linear power exceeds a certain threshold which may be determined as a function of the characteristics of the fuel.
To date, the techniques of protection of pressurized-water nuclear reactors against the appearance of one of these two phenomena have consisted in calculating parameters representative of these phenomena and triggering an emergency shutdown when one of the parameters exceeds a limiting value which has been assigned to it, this limiting value taking into account certain inaccuracies in the measurements, the more or less representative character of the parameter and the accuracy of evaluation of these parameters.
Hence one may be led to trigger emergency shutdowns in cases which are not really critical if the determination of the parameters is done with a certain inaccuracy.
The parameters representative of the crisis of boiling and the linear power may be determined by calculation, from a measurement of the inlet and outlet temperatures of the coolant in the reactor, of the pressure and flow of this coolant, and from the distribution of power in the core of the reactor.
In order to increase the accuracy of the determination, one may obviously increase the accuracy of the measurement of temperature pressure and flow, but to date the distribution of power in the core of the reactor could not be determined with accuracy during operation of the nuclear reactor, and one was satisfied to employ envelope values for this distribuiton, whether it was axial distribution or radial distribution of power.
In the methods commonly employed for handling pressurized-water nuclear reactors, the axial deviation of the flux is preferably maintained at a constant value.
The axial deviation of the flux is a parameter representative of the imbalance between the neutron flux from the upper portion of the core .phi.H and the neutron flux from the lower portion of the core .phi.B. The parameter is equal to the ratio ##EQU1##
Hence in this method of handling at constant axial deviation the only parameter measured as far as the distribution of flux or power in the core is concerned is the axial deviation of the flux.
However, for one and the same value of axial deviation, it is possible to conceive forms of distribution of the flux along the fuel element which are extremely variable.
In order to take into account this uncertainty about the distribution, for the calculation of the parameters relative to the crisis of boiling and the linear power, an extremely unfavorable distribution is employed, called the envelope distribution, which enables all of the most favorable cases of distribution to be included, which would lead to one and the same axial deviation of the flux.
As far as the radial distribution of the flux is concerned one takes into account an envelope factor of distribution which takes into account every possible situation and is a function of the position of the control rods.
Hence such envelope distributions lead to a relatively penalizing evaluation of the parameters which can lead to emergency shutdowns in less than critical cases.
On the other hand, in methods of controlling the reactor which are different from the constant axial deviation methods employed conventionally to date, such as the method described in applicants' French Patent Application No. 77-19316, which employs groups of control rods which are less absorbent than the groups employed in the case of handling at constant axial deviation, for example, and an absorbent regulation group, it is no longer possible to employ an envelope value which takes into account the positions of the control rods for the radial distribution, because this envelope value would have to take into account both the positions of the not very absorbent groups for power recovery and the position of the regulation group, which would be extremely complicated, or solely the position of the power recovery groups, which would lead to a very approximate evaluation of the parameters. In this way, one could not totally profit by the flexibility of use of the reactor available with this method of handling.